The invention relates to a composite neutron absorbent material and to a process for manufacturing this material.
Neutron absorbent materials are neutron absorbers. They find application in the manufacture of control rods for example used to control the reactivity of nuclear reactors, in particular to control pressurized water nuclear reactors (PWR) and fast neutron reactors (FNR).
Inside the cores of nuclear reactors, neutron absorbent materials are indispensable components of the control rods. The latter form the command, adjustment and stoppage systems of reactor reactivity.
The materials which make up these control rods contain nuclides able to absorb the neutrons in order to reduce the reactor""s neutron flow.
These neutron-absorbing materials may for example be used in the two main types of French nuclear reactors: Pressurized Water Reactors and Fast Neutron reactors.
The first choice that a core designer must face is which neutron-absorbent nuclide to use. It must meet the requirements of anti-reactivity related to the energy rating of the core: type of fuel and fuel assembly, desired neutron flow, nuclear station power, etc. Depending upon desired core power, upon the intensity of the neutron flow, it is essential to provide the necessary antireactivity for normal adjustment of this neutron flow or possibly for the emergency stoppage of nuclear fission reactions. During the fission of a heavy nucleus in the core of a nuclear reactor, a few neutrons are released in the free state. If, among these neutrons that are released, some happen to meet a fissile nucleus and cause its fission they in turn generate descendants which themselves may cause the fission of another nucleus and give birth from generation to generation to a chain reaction. It therefore appears important to control the quantity of free neutrons formed in order to prevent the fission reaction from racing out of control and to maintain this fission in a critical state, that is to say in equilibrium.
Therefore the control rods containing the neutron absorbent materials are mobile rods mounted in the core of nuclear reactors such that they can slide between the fuel assemblies, or mounted in a network of fuel pins of an assembly. Control of core fission is made by inserting or withdrawing these rods from the core of the nuclear reactor by sliding them in or out of position.
The absorbent materials may be used to maintain nuclear fission in the critical state, in which case they form piloting rods. They may also be used to ensure quick stoppage of a chain reaction in which case they form safety rods.
Other criteria may be taken into account in the choice of absorbent material. These criteria are:
good mechanical properties, in particular a Young""s modulus that is as low as possible, moderate ultimate stress, good resistance to mechanical damage and more particularly good resistance to crack propagation,
a reasonable overall cost (raw material and manufacture),
good chemical and shrinkage resistance to radiation
optionally, chemical compatibility with the cladding (generally in stainless steel) which may be used to shield the absorbent.
A great amount of research had been conducted on boron carbide B4C with the sole view to its use as neutron absorber given its high effective neutron capture cross-section. B4C absorbent material is used in the form of stacks of sintered cylindrical pellets, made from powders.
Although having substantial chemical inertia, B4C oxides easily on and after 600xc2x0 C. in the presence of oxygen. This compound is also sensitive to water corrosion in the primary PWR medium, in particular when radiated by the neutrons or when subjected to neutron radiation. This is one of the reasons why it is generally inserted in stainless steel cladding.
Also, the lifetime of boron carbide never reaches the theoretical limit fixed by boron exhaustion on account of damage to the material caused by the large quantity of helium and lithium formed by neutron absorption 10B(n,xcex1), Li. Therefore, under the effect of temperature, one fraction of the helium formed diffuses outside the material while the other accumulates therein, causing swelling and micro-fracturing of the material.
In the particular case of fast neutron reactors, the flow of neutrons (energy greater than 1 MeV) penetrates inside all the absorbent material which causes a volume release of heat whereas the surface of the absorbent is directly cooled by sodium. A substantial radial thermal gradient therefore occurs in the material, which may reach several hundred degrees per centimeter. This thermal gradient involves major tangential heat stresses in the material which cause critical radial cracking and complete fragmentation of the absorbent material.
In the particular case of pressurized water reactors, the flow of neutrons (energy less than a few eV) only penetrates the peripheral part of the absorbent material. The swelling described previously therefore only occurs in a peripheral ring of the absorbent. Differential swelling therefore occurs between the circumference and the core of the absorber pellets which causes major radial stresses, critical tangential cracking superimposing itself upon the micro-cracking previously described, and complete fragmentation of the absorbent material.
The combination of swelling, microfracturing and cracking of the material may, under strong radiation, cause a mechanical interaction between the absorbent material and the steel cladding which may lead to fracture of the cladding which itself is subject to weakening firstly by fast neutron radiation and secondly by the diffusion of a certain amount of boron and carbon derived from the absorbent material.
These two modes of critical cracking are macroscopic crack phenomena with imposed strain.
It is therefore necessary to develop a neutron absorbent material which may be used in these two types of reactors.
This material therefore, in addition to the above-mentioned properties of a low Young""s modulus and a low coefficient of thermal expansion, must offer high heat conductivity, toughness, resistance to crack propagation and resistance to mechanical damage.
Document EP-A-0 359 683 describes a neutron absorbent pellet and its process of manufacture. It describes a scarcely absorbent element obtained by moulding or sintering a mixture of ceramic powders (B4C, HfO2, Eu2O3) and metal (Hf, Eu, Ni, Cr).
Document WO-A-94/28556 describes a neutron absorbent material and its method of preparation. The material described contains boron carbide and may contain hafnium, in particular hafnium diboride. The hafnium diboride represents no more than 40% by volume, preferably from 20 to 30% by volume.
The purpose of the present invention is precisely to remedy the above-mentioned disadvantages and to provide a neutron absorbent material having all the required properties, in particular for its use in control rods for a nuclear reactor.
The neutron absorbent material of the present invention is characterized in that it contains boron carbide and hafnium, in particular it may contain boron carbide and hafnium diboride.
According to the invention, the boron may account for at least approximately 65% by atoms of the material, for example approximately 72% by atoms of the material.
According to the invention, the hafnium may account for up to approximately 18% by atoms of the material, for example approximately 10% by atoms of the material.
According to the invention, the boron carbide may be in the form of particles having a diameter of up to approximately 50 xcexcm.
According to the invention, the hafnium may be in the form of agglomerates of hafnium boride whose size preferably ranges up to approximately 500 xcexcm, for example up to approximately 250 xcexcm.
The material of the present invention may have a density of approximately 2870 to 6800 kg/m3, for example of approximately 3220 to 5770 kg/m3, for example of 5165 kg/m3 or 5060 kg/m3.
The material of the present invention may also contain carbon and/or traces of hafnium carbide.
The material of the present invention may also contain zirconium representing up to approximately 4% by weight of the hafnium fraction. The zirconium may be an impurity present in the hafnium, and may represent up to 4% by weight of the hafnium.
The present invention also relates to a control rod for a nuclear reactor containing a neutron absorbent material according to the present invention.
The control rods may be command, adjustment and reactivity stoppage rods for nuclear reactors. The neutron absorbent material of the present invention may, for example, be used in all the above-mentioned applications.
The present invention also concerns a process for manufacturing a neutron absorbent material, said neutron absorbent material being a composite material containing boron carbide and hafnium diboride. The boron carbide and hafnium diboride may be the main components of the material of the invention.
This process comprises the following steps:
adding hafnium powder to a powder of boron carbide,
mixing the boron carbide powder and the hafnium powder in homogeneous manner, and
sintering the homogeneous mixture obtained at a sintering pressure and temperature that are sufficient to obtain the composite material.
According to the process of the invention, between 15 and 40% by volume of hafnium may be added, for example between 20 and 30% by volume of hafnium, for example around 25% by volume of hafnium. It is also to be noted at this point that the hafnium may contain a weight fraction of zirconium of up to 4% by weight of the hafnium, the zirconium possibly being an impurity usually found in hafnium. The homogeneous mixture of the powders of boron carbide and hafnium, represents 100% by volume of the mixture.
According to the process of the invention, the boron carbide powder may preferably have a grain size of up to approximately 50 xcexcm.
According to the process of the invention, the hafnium powder may preferably have a grain size of up to approximately 20 xcexcm, further preferably of up to approximately 10 xcexcn.
According to the process of the invention, the mixture of boron carbide powder and hafnium powder may be made using any means known to persons skilled in the art to obtain a homogeneous mixture of said powders. Preferably, the mixture of these powders may be made by simultaneous application of mechanical stirring and ultrasound, or simply by application of ultrasound to a paste containing said powders dispersed in a dispersion liquid.
The dispersion liquid is preferably a liquid which, after the mixing of the powders, may be easily removed by evaporation for example. This dispersion liquid may, for example, be an alcohol such as ethanol.
When a homogeneous mixture of the powders is obtained, it may be dried by alcohol evaporation for example, and then screened in order to remove any powder aggregates in the mixture.
The homogeneous mixture obtained is then sintered to obtain the composite material.
According to the invention, sintering may be conducted in a vacuum or in any atmosphere made up of a neutral gas, preferably argon for example.
According to the invention, the mixture may be sintered in any appropriate mould for sintering such powders, for example a mould in graphite preferably lined with a sheet of graphite. Lining of the mould particularly prevents diffusion of the chemical species in the mixture towards the mould and facilitates subsequent release of the composite material from the mould.
The mould may be of appropriate shape to mould the material in pellet, plate, cruciform or pin shape and generally in a shape appropriate for forming control rods of a nuclear reactor.
Sintering of the mixture is advantageously conducted under conditions of temperature, pressure and time which enable proper densification of the composite material, it may for example be conducted at a temperature of approximately 1800 to 2100xc2x0 C., for example at a temperature of around 2000xc2x0 C., under a pressure of 70 to 110 MPa, for example from 90 to 100 MPa, for example of around 92 MPa, for a time of approximately 15 to 90 minutes, for example for approximately 1 hour.
Sintering may, for example, be conducted in a furnace maintained in a dynamic vacuum.
Sintering pressure may be applied before the reaction which occurs in the mixture of the boron carbide and hafnium powders. Therefore, the pressure may be applied before the temperature of the powder mixture reaches the sintering reaction temperature of said mixture, for example when the temperature of the mixture and the mould is approximately 20 to 1200xc2x0 C., for example around 500 to 1000xc2x0 C., or further at approximately 800xc2x0 C. for example.
Under these conditions, reactive sintering in the powder mixture, thermal contact between the boron carbide and hafnium powders and the composition, and the morphology of the composite material are accurate. In addition, the densification of the composite material of the present invention is most satisfactory.
One first advantage of the neutron absorbent material of the invention is that it offers greater resistance to crack propagation.
Another advantage of the material of the invention is that it maintains its geometrical intactness despite major mechanical damage. For the composite material of the invention does not fragment when subjected to a strong heat gradient, swelling, microfracturing or cracking.
A further advantage according to the invention was demonstrated by biaxial bending tests performed on sample disks of pure boron carbide material. These tests showed that the fracture behaviour of pure boron carbide was solely fragile, leading to fragmentation of the material. With the composite material of the invention, fracture behaviour is pseudo-plastic and dissipating: it does not lead to fragmentation of the composite material of the invention which, after mechanical damage, remains in a single block even if microfractured or cracked. This result indicates a Young""s modulus of the composite material of the present invention that is lower than that of pure boron carbide, to an elongation to fracture of the composite material of the present invention that is greater than that of pure boron carbide, and to a Weibull""s modulus of the composite material of the present invention that is greater than that of pure boron carbide, and hence indicates lesser dispersion of fracture probability in relation to a given stress which may occur in the material.
Another advantage of the invention was shown by comparative double-torsion tests on sample plates of pure boron carbide and a material of the invention. These tests showed pure boron carbide to have solely fragile crack behaviour with intragranular cracking. For the composite material of the invention, crack behaviour showed phenomena of repeated crack blocking, phenomena of crack deflection, and crack bridging phenomena in the composite structure of the material of the invention. The result was resistance to crack propagation and toughness of the composition material of the present invention that was greater than that of pure boron carbide. This result also shows an R curve effect exhibited by the composite material of the present invention through the double torsion test.
A further advantage of the invention was shown by heat gradient cracking tests on sample disks of pure boron carbide and sample disks of the material of the present invention. These tests showed solely fragile cracking and instantaneous intragranular cracking with the pure boron carbide material. Pure boron carbide cannot withstand a heat gradient representing the heat gradient which occurs under radiation in a fast neutron reactor. This is due to the thermo-mechanical stresses induced by the heat gradient which exceed the crack propagation resistance of pure boron carbide and causes fragmentation of the sample of pure boron carbide. With the composite material of the present invention, cracks also occurred in the sample disks but they were blocked by the structure of the material of the invention. The sample disks of the composite material of the present invention remained in a single piece. The cracks were shorter as their growth was blocked by the presence of agglomerates of hafnium boride which limited their propagation. This result is the effect of the heat conductivity, resistance to crack propagation and toughness of the composite material of the present invention which are greater than those of pure boron carbide, and of an R curve effect exhibited by the composite material of the present invention with the double-torsion test.
The composite absorbent material of the invention may, for example, be described as containing a homogeneous matrix of boron carbide (B4C) in which agglomerates are distributed in homogeneous manner which contain hafnium monoboride (HfB), hafnium diboride (HfB2) and traces of hafnium carbide (HfC). Zones of free carbon (C) can be detected on the interface between the matrix and the agglomerates, preferably this free carbon phase does not exceed 10% by weight of the boron carbide phase. The added hafnium therefore develops into hafnium boride phases present in the composite in the form of agglomerates.
The hafnium boride also offers the advantage of having very satisfactory reaction to neutron radiation which is better than that of boron carbide in terms of accommodation of strain, radiation defects and microfracturing.
The type of the agglomerate phases during reactive sintering may therefore be of importance in understanding the advantage of the composite material of the invention.
The phases of hafnium boride in the composition of the present invention have the appearance of agglomerates. These agglomerates in the composite material of the invention provide the composite material with the capacity to resist against crack propagation and to maintain its physical intactness despite the slow, progressive lowering of its mechanical characteristics.
Other characteristics and advantages will be further seen on reading the following examples which are evidently given for illustrative purposes and are not restrictive, with reference to the appended figures.